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Title: CTF: A modernized, production-level, thermal hydraulic solver for the solution of industry-relevant challenge problems in pressurized water reactors

Journal Article · · Nuclear Engineering and Design
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  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  2. Pennsylvania State Univ., University Park, PA (United States)
  3. Idaho National Lab. (INL), Idaho Falls, ID (United States)
  4. North Carolina State Univ., Raleigh, NC (United States)
  5. Westinghouse Electric Company LLC, Cranberry Township, PA (United States)
  6. AREVA NP, Courbevoie (France)
  7. Core Physics, Inc., Raleigh, NC (United States)
  8. Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
  9. M Pilch Consulting, Inc., Tijeras, NM (United States)
+ Show Author Affiliations

CTF is a thermal hydraulic (T/H) subchannel tool that has been extensively developed over the past ten years as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) program. The code was selected early in the CASL program for support of high-impact challenge problems that were found to be relevant to the nuclear industry and its currently operating fleet of pressurized water reactors (PWRs), including issues such as departure from nucleate boiling (DNB), crud-induced power shifts (CIPSs), and reactivity-insertion accidents (RIAs). By incorporating CTF into the multiphysics Virtual Environment for Reactor Application (VERA) core simulator software developed by CASL, CTF has become the primary means of providing fluid and fuel thermal feedback, as well as T/H figure-of-merits (FOMs) in large-scale reactor simulations. With the goal of solving industry challenge problems, CASL placed great emphasis on developing high-quality, high-performance, validated software tools that offer higher fidelity than what is currently possible with current industry methods. In support of this effort, CTF was developed from a research tool into an nuclear quality assurance (NQA-1)–compliant, production-level software tool that is capable of addressing the stated challenge problems and goals of CASL. This work presents a review of the major technological achievements that were realized in developing CTF over the past decade of the CASL program and presents an overview of the code solution approach and closure models.

Research Organization:
Idaho National Laboratory (INL), Idaho Falls, ID (United States); Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE)
Grant/Contract Number:
AC05-00OR22725; AC07-05ID14517
OSTI ID:
1885341
Journal Information:
Nuclear Engineering and Design, Journal Name: Nuclear Engineering and Design Journal Issue: 1 Vol. 397; ISSN 0029-5493
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English

References (21)

Explicit approximations to the solution of Colebrook's friction factor equation journal May 1982
Pressurized water reactor fuel crud and corrosion modeling journal August 2011
Prediction of departure from nucleate boiling for an axially non-uniform heat flux distribution journal January 1967
Optimization and parallelization of the thermal–hydraulic subchannel code CTF for high-fidelity multi-physics applications journal October 2015
Implementation and assessment of wall friction models for LWR core analysis journal May 2018
The Virtual Environment for Reactor Applications (VERA): Design and architecture journal December 2016
Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event journal December 2016
A new fuel modeling capability, CTFFuel, with a case study on the fuel thermal conductivity degradation journal January 2019
Development and implementation of a CTF code verification suite journal December 2020
The turbulence structure of equilibrium boundary layers journal September 1967
An Improved Energy Deposition Model in Mpact and Explicit heat Generation Coupling with ctf journal January 2021
Implementation of a Spacer Grid Rod Thermal-Hydraulic Reconstruction (ROTHCON) Capability into the Thermal-Hydraulic Subchannel Code CTF journal April 2019
BISON: A Flexible Code for Advanced Simulation of the Performance of Multiple Nuclear Fuel Forms journal March 2021
Flow Boiling Heat Transfer in Vertical Tubes Correlated by an Asymptotic Model journal January 1992
Letter to the Editor journal September 2019
Onset of Nucleate Boiling and Active Nucleation Site Density During Subcooled Flow Boiling journal July 2002
Heat-Transfer Augmentation in Rod Bundles Near Grid Spacers journal February 1982
VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB
  • Kucukboyaci, Vefa N.; Sung, Yixing; Xu, Yiban
  • Volume 4: Computational Fluid Dynamics (CFD) and Coupled Codes; Decontamination and Decommissioning, Radiation Protection, Shielding, and Waste Management; Workforce Development, Nuclear Education and Public Acceptance; Mitigation Strategies for Beyond Design Basis Events; Risk Management https://doi.org/10.1115/ICONE24-60865
conference June 2016
COBRA-TF Simulation of Fuel Thermal Response During Reactivity Initiated Accidents Using the NSRR Pulse Irradiation Experiments conference November 2014
VERA Core Simulator Methodology for Pressurized Water Reactor Cycle Depletion journal January 2017
Implementation of a New Wall Boiling Model in CTF conference January 2020

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